Process for preparing multi-component nuclear fuels

ABSTRACT

A process for preparing particles, powders, spheres, and other shapes of uranium carbide or uranium nitride and/or thorium carbide or thorium nitride containing significant amounts of a second fissionable component. The particles are prepared by impregnating a matrix with a sol or other suitable dispersion of the fissionable component.

United States Patent Triggiani et al.

[451 June 20, 1972 PROCESS FOR PREPARING MULTI- COMPONENT NUCLEAR FUELS[72] Inventors: Leonard V. Triggianl, Rockville; Moises G. Sanchez,Sevema Park, both of Md,

[73] Assignee: W. R. Grace 8: Co., New York, NY.

[22 Filed: Sept. 24, 1960 2| Appl. No.: 860,814

Related U.S. Application Data [63] Continuation-in-part of Ser. No.670,394, Sept. 25,

1967, Pat. No. 3,514,412.

[52] U.S. Cl. ..252/30l.l S, 23/347, 23/349, 176/68 [51 rm. Cl. ..C09k3/00 Primary Examiner-Carl D. Quarforth Assistant ExaminerR. L. TateAttorney-Joseph P. Nigon and Kenneth E. Prince ABSTRACT A process forpreparing particles, powders, spheres, and other shapes of uraniumcarbide or uranium nitride and/or thorium carbide or thorium nitridecontaining significant amounts of a second fissionable component. Theparticles are prepared by impregnating a matrix with a sol or othersuitable dispersion of the fissionable component.

10 Claims, No Drawings PROCESS FOR PREPARING MULTl-COMPONENT NUCLEARFUELS This application is a continuation-in-part of application serialnumber 670,394, filed Sept. 25, i967, now U.S. Pat. No. 3,514,412.

The preparation of fuel elements from sols has resulted in products thathave very desirable physical properties. The products can be sintered tovery high density at much lower temperatures than was possible when thefuel elements were prepared by conventional ceramic techniques.Microspheres prepared from sols may range in size from a few microns upto l ,000 microns or more and provide a very convenient method ofhandling nuclear fuels.

There has been demand for fuel elements of binary and multi-componentstructure containing more than one material. Plutonia-urania fuels, forexample, are in demand as are the thoria and urania fuels containingfissionable materials such as U-235 or U-233.

In the prior art processes, these compositions have been prepared byco-precipitation from solutions of salts of the respective metals or byphysical mixing of the dried oxide particles followed by comminution,pressing, etc., by means of conventional ceramic techniques. They havealso been prepared from mixed oxide sols in a process wherein a solutionor sol of the second component is added to a sol of the first component.

The differences in the level of the radio-toxicity of the fertile matrixcomponent and the fissionable second component increases the problems inpreparing mixed oxide sol products by these conventional techniques.Urania U-238) can be handled safely with conventional laboratoryequipment. Plutonium, on the other hand, can only be handled in gloveboxes and other sophisticated equipment. The handling problems aremultiplied several fold when the mixed oxide fuel preparation is scaledup for plant production.

We have developed a process for preparing these mixed fuel systems inwhich the two components are handled separately in equipment designed tohandle each component. This system is a substantial improvement over theconventional processing in which both components are processed togetherin each of the steps, so that the most sophisticated equipment isnecessary for each step and for the entire quantity of material beingprocessed.

In our novel process, we impregnated porous matrix particles withcarbide or nitride precursors of the fissionable component. This may beaccomplished by impregnating the spheres with sols or solutions of asalt of the desired fissionable component followed by conversion to thecarbide or nitride in the pores ofthe matrix.

The particles may be subjected to various degrees of drying. The spheresmay properly be termed gels in microspheroidal form. The term "gel isthus applicable to describe our matrix materials.

Our process fills a long standing need for means of preparing a mixedcarbide or nitride fuel wherein a fissionable component is isolated toonly a few processing steps, so that the greatest amount of work can bedone in conventional equip ment. This system also avoids the problem ofcontamination of the equipment with the fissionable materials prior tothe final step of the process.

For purposes of simplicity, our process will be described as ourpreferred microsphere impregnation process. However, it is obvious thatour process can be used for preparing nuclear fuel particles in anydesired shape or physical form.

In our preferred process, we prepare the fertile matrix microspheres(thoria or urania mixed with carbon) and add the desired amount of thefissionable component as a solution of a salt or as a sol of thecomponent. The impregnated microspheres are then washed, dried, andcalcined to prepare the final product. The nitride is prepared bycalcining a urania plus carbon, thoria plus carbon or thoria-urania pluscarbon mixture in an argon atmosphere followed by calcination in anitrogen atmosphere under conditions suitable for conversion to thenitride. The fissionable component is normally added as a solution of asalt or a sol of the fertile microspheres prior to the calcinationsteps.

Broadly speaking, our preferred carbide process comprises the followingsteps:

1. Selection and dissolution of the fuel raw materials.

2. Addition of a suitable quantity of carbon to convert the fertile andfissionable components to the carbides.

3. Preparation of sols or suitably modified solutions of thesematerials.

4. Formation of microspheres from the carbon containing sols or thesuitably modified carbon containing solutions.

5. Addition of a fissionable additive into the microsphere product.

6. Sintering the carbon containing microspheres to efi'ect conversion tothe carbide.

Our preferred nitride preparation process comprises the following steps:

1. Selection and dissolution of the fuel raw materials.

2. Preparation of sols or suitable modified solutions of thesematerials.

3. Addition of a sufficient quantity of carbon to convert thefissionable and fertile materials to the carbide.

4. Formation of microspheres from the carbon containing sols or thesuitable modified carbon containing solutions.

5. Addition of the fissionable additive into the microsphere product.

6. Calcination of the carbon containing microspheres in an argonatmosphere to form the carbide.

7. Conversion of the carbide to the nitride by calcination in a nitrogenatmosphere.

In the first step of our process, the fertile base material and thefissionable materials are selected. The fertile matrix materials arenaturally occuring or depleted urania (U-238) mixed with carbon above orin admixture with other materials such as thoria, for example, Thoria(Th-232) mixed with carbon may also be used as a fertile matrixmaterial.

The fissionable second component of the microspheres may be uranium orplutonium, for example. Urania U-238) mixed with carbon, for example,may be impregnated with plutonium or the other fissionable isotopesU-233 or U-235. A mixed thorium-uranium carbide fuel may be prepared byimpregnating thoria mixed with carbon (ThO -C) microspheres with U- 233or N-235.

The fertile matrix material, urania or thoria, is first obtained as asolution of nitrate, chloride, etc. The solution is then converted tothe sol form and mixed with carbon. Suitable sols may be prepared by anyof several methods. The preferred techniques for sol formation are:

l. Electrodialysis using anion permeable membranes.

2. Controlled hydrolysis with urea.

3. ion exchange using resin in the hydroxide form.

4. Peptization of washed hydroxides with an acid.

5. Electrolysis of solutions, with oxidation of the anions to a volatilecomponent.

in the next step of our process, the sols are converted to microspheres.The method of preparing these microspheres is not part of thisinvention. it is described in US. Pat. No. 3,331,785. Briefly, theprocess comprises forming the sols into droplets and drying the dropletsin a column of solvent passed in countercurrent direction to the solparticles. The formed microspheres are removed from the bottom of thecolumn and washed.

The matrix material may also be prepared as a powder or as microspheresor as larger sized spheroids in a process in which a solution of a saltof the matrix material is admixed with a water soluble resin thatincreases in viscosity in an alkaline medium. The droplets of solutionare then fed into an aqueous alkaline solution to form microspheres orspheroids. The particles or spheroids are recovered, washed, and dried.

The microspheres or particles may also be prepared by any of the otherprocesses described in the technical and patent literature, providedthat the final product has sufficient porosity to retain the desiredamount of the fissionable component.

In our process. we believe that the thoria-carbon and urania-carbonmicrospheres when contacted with the solution or so] containing thefissionable component accommodates the fissionable component solutionwithin the voids of the microspheres. The solution may then by convertedto an insoluble form, dried and sintered to form the carbide or nitrideproduct.

The solution or sol of the fissionable material used to impregnate themicrospheres may be prepared in an inorganic or organic solvent. Anaqueous solution of the salt is preferred, since it is thereby possibleto achieve a higher concentration of fissionable solution. The preferredsalt is the nitrate; however. the chloride, sulfate. etc.. can also beused in the preparation of the impregnation solution. The impregnationsolution of fissionable material is prepared in concentrations of about0.1 to 700 grams per liter. When an organic solvent is used. thepreferred solvent is acetone. Other suitable solvents include diethylether. dibutyl ether. methylisobutylketone, tributylphosphate.trioctylamine, trilaurylamine, cyclohexyldilaurylamine, certainalcohols, etc.

When the fertile base microspheres are composed of urania, admixed withcarbon. they may be hyperstoichiometric in oxygen at this stage. due tothe presence of hexavalent uranium. Hexavalent uranium is more solublein aqueous media then quadrivalent uranium. In that case. themicrospheres must be reduced to the dioxide if the product is to be freeof interparticle sludge. etc. and to insure proper interaction of theimpregnant and urania substrate. This reduction can by carried out usingany suitable technique. such as hydrogen reduction, etc. However. thisstep can by omitted if a non-aqueous solvent is used to prepare theimpregnant. The hydrogen treatment is normally carried out at atemperature of 300-900C. for about 1 hour to hours. In our process. thisstep may be completed at low temperatures without destruction of themicrosphere structure or porosity.

In one acceptable technique. the spheres were impregnated by slowlyadding a solution of the fissionable salt or the sol to the microsphereswhile they are being agitated. When the fertile matrix material isurania-carbon. a portion of the agitation is preferably provided by aflow of argon or other inert gas that prevents the oxidation of uraniato the hexavalent state. Any suitable inert gas. such as helium. argon.neon, nitrogen. etc.. may be used. In the laboratory. it is convenientto agitate the microspheres by regulating the flow of gas in the areasurrounding the microspheres. Other mechanical techniques for agitationmay be used, such as stirrin shaking, etc. A itation ing themicrospheres in an aqueous ammonia solution. We prefer concentratedaqueous ammonia. However. concentrations of ammonia between 5 and 30weight percent may be used. Gaseous NH has also been used for thispurpose. Generally, the precipitation is complete in about 10 minutes.However, shorter or longer times may be dictated by the type ofoperation and equipment being used.

The microspheres are then washed in deionized water to remove excessammonia. anions, hexanol or other solvent. Generally about 250 cc. ofdeionized water per gram microspheres is sufficient to remove allimpurities. The microspheres are then vacuum dried.

The sintering step is the final step of our process. Sintering ispreferably carried out in a hydrogen-nitrogen atmosphere by heating at300700C. for about 0.5 to 7 hours, followed by sintering for another 0.5to 6 hours at 1,000C. to 1.800C.

The matrix particles may. of course, be separated according to size byscreening or other technique and only those particles falling in a givensize range impregnated with the fissionable materials.

Our invention is further illustrated by the following specific butnon-limiting examples.

EXAMPLE I This example illustrates a suitable method of preparingurania-carbon microspheres for use as a fertile base.

A uranous chloride solution containing 7.7 weight percent uranium havinga density of 1.14 grams per cc. was prepared from a uranic oxychlorideUO Cl solution by reduction with uranium metal. A total of 500 cc. ofthis solution was added with vigorous agitation to an alkalinedispersion of carbon. The carbon dispersion was a commercially availableproduct containing 22 weight percent carbon. A total of 43.5 grams ofthe commercial carbon dispersion was diluted with 400 ml. of waterbefore addition of the uranous chloride solution and a total of 250 cc.of concentrated ammonium hydroxide solution was added over a period of 1hour. At the end of this time. a dark precipitate had collected in thebottom of the reaction vessel. The precipitate was then washed withdilute aqueous ammonia and water and slurried in a small amount ofwater. Nitric acid was added to a pH of 3.4 and the slurry was peptized1% hours at C. The resulting sol was urania coated carbon spheres havinga theoretical C/U ratio of 5.57.

The final H of the sol was corrected to a value of 3.8 b addi- In thisrun, a 1 gram sample of urania-carbon microspheres was pretreated byheating the urania-carbon microspheres to a temperature of 500C. for 3hours in hydrogen. This step removed any excess oxygen and converted theurania to stoichiometric U A 1 gram sample of these microspheres wasplaced in a fritted disc filtration funnel in a glove box. A flow ofargon was maintained through the funnel to blanket the spheres in anargon atmosphere during the impregnation step. A stock plutoniumsolution about 6 molar in nitric acid was prepared containing 60 gramsof plutonium per liter. 2 ml. of this stock solution was diluted with 2ml. of deionized water to prepare the impregnation solution. Thissolution was added dropwise to the microspheres. The microspheres wereconstantly agitated during the addition. Agitation was provided by aflow of argon through the filtration funnel. A total of about 0.8 cc. ofsolution was added by a medicine dropper to bring the spheres toincipient wetness.

The spheres were then washed with approximately 30 cc. of dry hexanol toremove water. The hexanol was removed very gently by vacuum filtrationso as not to damage the impregnated microspheres.

The sample was then ammoniated by contacting with about 90 cc. ofconcentrated ammonia. The contact time was 15 minutes. The ammonia wasthen removed very gently by vacuum filtration. No precipitate wasobserved in the ammonia filtrate. The spheres at this point were wholewith no precipitate on the surfaces. The impregnated microspheres werethen washed with deionized water for 4 hours during which time nophysical change in the microspheres was observed.

The spheres were dried using conventional techniques and after dryingwere treated in hydrogen with 1 hour required to reach 500C. Thistemperature was maintained for 3 hours. The furnace was evacuated andthe temperature raised over a period of 'r hour to 1,150C. and held for3 hours. At the end of 3 hours, the pressure was 5 X millimeters ofmercury. The temperature was raised over a period of minutes to l.35()C.at a rate such that the pressure did not exceed 5 X 10 millimeters ofmercury. The temperature was then raised to 1,400C. and held for 1 hourto achieve a pressure of 7 X l0 millimeters of mercury. The finaltemperature, 1,750C., was achieved very rapidly and held for 3 hours.The ultimate vacuum achieved was in the 10 millimeter range.

The product had a density of 11.2. This compares with a theoreticaldensity of pure UC of 13.63 and pure UC of l 1.68. Analysis by X-ray ofthe product microspheres showed a mole ratio of UC. to UC to be 52 to48. The total carbon content ofthe sintered spheres was 7.28 weightpercent.

EXAMPLE 3 This example illustrates the method of converting theimpregnated urania-carbon microspheres to the nitride.

Approximately 50 grams of impregnated uranium oxidecarbon microspheres.prepared by the general technique described in example 1, having acarbon to uranium molar ratio of 2.37 were transferred to graphitecrucible. The crucible was inserted in a ceramic tube between thesilicon-carbide heating elements of a Burrell furnace. Hydrogen gas flowwas started through the furnace. The temperature increased to 500C. overa period of 1 hour. The temperature was maintained at 500C. for 3 hours.The gas feed was changed to argon. The temperature was increased to1,500C. over a 3% hour period. The gas feed was changed to argon andnitrogen sweep continued at a temperature of 1,400C. for a period of 7hours. The temperature was again increased to 1,500C. for an additionalperiod of 7 hour. The product was removed, analyzed, and found tocontain 069 percent free carbon and 5 80 parts per million of oxygen.

An effort was made to increase the theoretical density and decrease theoxygen content to the impregnated UN product. The product wastransferred to a commercially available high temperature furnaceequipped with means for sweeping gas to the furnace. A sample wasbrought to a temperature of 1,700C. in an argon atmosphere and heldthere for about 24 hours. The furnace was then cooled to roomtemperature and the product analyzed. The product was foundto contain410 parts per million in oxygen and have a density on the order ofpercent of theoretical.

EXAMPLE 4 When plutonium is precipitated in the pores of a gelsubstrate, the skeletal thickness may be thought of as a fertile matrixbarrier separating the plutonium precipitated into the pores. Onsintering, the plutonium will diffuse through the skeletal wall. We maytherefore conclude that, on an average, about one-half of the skeletalwall thickness will represent the maximum diffusion path necessary forthe plutonium (or other fissionable component) and the fertile matrixmaterial to achieve homogeneous solid solution.

It is quite clear that gel material having thin skeletons are desirableas impregnation substrates. A series of calculations were made to definethese properties of our matrices.

The surface area and pore volumes of a series of thoria sol residuesprepared by the electrodialysis technique were measured. In this series,the surface areas varied from 85 square meters per gram to 127 squaremeters per gram; the pore volumes from 0.1 l to 0.13 cc. per gram.

Using these figures and takingthe density of thoria at 10 grams percubic centimeter, we are able to calculate the specific skeletal volume,that is the volume occupied by the metal oxide comprising the particleframework. We find this value is 0.10 cc. per gram.

The pore volume is equal to the void space per gram. The specific volumeis then equal to:

pore volume plus skeletal volume, or in this case (where the pore volumeequals 0.13) to 0.23.

The porosity can be calculated using the formula:

Pore volume (cm/per gram) Total volume (em/per gram) 0 Volume 2" Surfacearea Using the surface area and pore volume data referred to above, wefound our matrix materials have a skeletal thickness (0) offrom 7.9 to118 A.

A basic characteristic of our materials is a combination of surface areaand pore volume which will minimize the diffusion paths of less than 500A are required for homogeneous solid solution formation. Of course, theshorter the diffusion path, the easier the attainment of solid solution.ln our process, we prefer to use materials with diffusion paths of lessthan 500 A., preferably less than A.

Our fertile base materials can thus be characterized as havmg:

Porosities of 10 to 80 percent:

An average diffusion path of less than 500 A., preferably less than 100A.

A fissionable material content of l to 25 weight percent.

What is claimed is:

l. A process for preparing a mixed carbide nuclear fuel comprising afertile matrix selected from the group consisting of uranium 238carbide, thorium 232 carbide and mixtures thereof and a fissionablecomponent selected from the group consisting of plutonium 239 carbide,uranium 233 carbide and uranium 235 carbide and mixtures thereof whichcomprises:

a. mixing carbon with a material selected from the group consisting ofurania 238 sols, thoria 232 sols thoria 232- urania 238 sols, solutionsof uranium 238 compounds, solutions of thorium 232 compounds andsolutions of thoria 232-urania 238 compounds,

b. drying said sols or solutions and forming urania 238-C,

thoria 232-C, or thoria 232-urania 238-C gel particles,

c. contacting said particles with a fluid containing a compound offissionable material selected from the group consisting of plutonium239, uranium 233 and uranium 235,

d. calcining said particles under vacuum or in an inert atmosphere at atemperature of about l,000 to l,900C. for about 3 to 30 hours, and

e. recovering the mixed carbide nuclear fuel particles as a product.

2. The process according to claim 1 wherein the impregnated particlescontain from about 1 to 30 percent fissionable material.

3. The process according to claim 1 wherein the amount of carbon mixedwith the urania sol or solution is about percent in excess of thestoichiometric amount required to form the dicarbide, and theimpregnated particles are calcined at a temperature of about 1, l 50 to1,450C. for about 4 to 6 hours, followed by heating to about l,850C. for3 to 4 hours to densify the particles.

4. The process according to claim 1 wherein the urania 238 particles areheated in hydrogen or other reducing atmosphere at a temperature ofabout 400-700C. for a period of about 0.5 to 3 hours.

5. A process for preparing a mixed nitride, nuclear fuel comprising afertile matrix selected from the group consisting of uranium 238nitride, thorium 232 nitride and mixtures thereof and a fissionablecomponent selected from the group consisting of plutonium 239 nitride,uranium 233 nitride and uranium 235 nitride and mixtures thereof, whichcomprises:

a. mixing carbon with a material selected from the group consisting ofurania 238 sols, thoria 232 sols thoria 232- urania 238 sols, solutionsof uranium 238 compounds, solutions of thorium 232 compounds andsolutions of thoria 232-urania 238 compounds,

b. drying said sols or solutions and forming urania 238-C,

thoria 232-C, or thoria 232-urania 238-C gel particles,

c. contacting said particles with a fluid containing a compound of afissionable material selected from the group consisting of plutonium239, uranium 233 and uranium 235,

d. calcining said particles in a nitrogen atmosphere at a temperature of1,000 to l,900C. for about 0.5 to l0 hours, and

e. recovering the mixed nitride nuclear fuel product.

6. A process for preparing a mixed nitride, nuclear fuel comprising afertile matrix selected from the group consisting of uranium 238nitride, thorium 232 nitride and mixtures thereof and a fissionablecomponent selected from the group consisting of plutonium 239 nitride,uranium 233 nitride and uranium 235 nitride and mixtures thereof, whichcomprises:

a. mixing carbon with a material selected from the group consisting ofurania 238 sols, thoria 232 sols thoria 232- urania 238 sols, solutionsof uranium 238 compounds, solutions of thorium 232 compounds andsolutions of thoria 232-urania 238 compounds,

b. drying said sols or solutions and forming urania 238-C,

thoria 232-C, or thoria 232-urania 238-C gel particles,

c. contacting said particles with a fluid containing a compound of afissionable material, selected from the group consisting of plutonium239, uranium 233 and uranium 235, d. calcining in an argon atmosphere ata temperature of about 1,000 to l,900C. for about 3.5 to 30 hours,followed by e. calcining in a nitrogen atmosphere at a temperature ofl,000 to l,900C. for about 6 to 10 hours, and

f. recovering the mixed nitride nuclear fuel product.

7. The process according to claim 6 wherein the impregnated particlescontain about 1 to 30 percent fissionable materials.

8. The process according to claim 6 wherein the particles are calcinedin each of the calcination steps for a period of about 24 hours.

9. The process according to claim 5 wherein the urania 238 particles areheated in hydrogen or other reducing atmosphere at a temperature ofabout 400700C. for a period of 0.5 to 3 hours.

10. The process according to claim 6 wherein the urania 238 particlesare heated in hydrogen or other reducing atmosphere at a temperature ofabout 400-700C. for a period of0.5 to 3 hours.

2. The process according to claim 1 wherein the impregnated particlescontain from about 1 to 30 percent fissionable material.
 3. The processaccording to claim 1 wherein the amount of carbon mixed with the uraniasol or solution is about 20 percent in excess of the stoichiometricamount required to form the dicarbide, and the impregnated particles arecalcined at a temperature of about 1,150* to 1,450*C. for about 4 to 6hours, followed by heating to about 1,850*C. for 3 to 4 hours to densifythe particles.
 4. The process according to claim 1 wherein the urania238 particles are heated in hydrogen or other reducing atmosphere at atemperature of about 400*-700*C. for a period of about 0.5 to 3 hours.5. A process for preparing a mixed nitride, nuclear fuel comprising afertile matrix selected from the group consisting of uranium 238nitride, thorium 232 nitride and mixtures thereof and a fissionablecomponent selected from the group consisting of plutonium 239 nitride,uranium 233 nitriDe and uranium 235 nitride and mixtures thereof, whichcomprises: a. mixing carbon with a material selected from the groupconsisting of urania 238 sols, thoria 232 sols thoria 232-urania 238sols, solutions of uranium 238 compounds, solutions of thorium 232compounds and solutions of thoria 232-urania 238 compounds, b. dryingsaid sols or solutions and forming urania 238-C, thoria 232-C, or thoria232-urania 238-C gel particles, c. contacting said particles with afluid containing a compound of a fissionable material selected from thegroup consisting of plutonium 239, uranium 233 and uranium 235, d.calcining said particles in a nitrogen atmosphere at a temperature of1,000* to 1,900*C. for about 0.5 to 10 hours, and e. recovering themixed nitride nuclear fuel product.
 6. A process for preparing a mixednitride, nuclear fuel comprising a fertile matrix selected from thegroup consisting of uranium 238 nitride, thorium 232 nitride andmixtures thereof and a fissionable component selected from the groupconsisting of plutonium 239 nitride, uranium 233 nitride and uranium 235nitride and mixtures thereof, which comprises: a. mixing carbon with amaterial selected from the group consisting of urania 238 sols, thoria232 sols thoria 232-urania 238 sols, solutions of uranium 238 compounds,solutions of thorium 232 compounds and solutions of thoria 232-urania238 compounds, b. drying said sols or solutions and forming urania238-C, thoria 232-C, or thoria 232-urania 238-C gel particles, c.contacting said particles with a fluid containing a compound of afissionable material, selected from the group consisting of plutonium239, uranium 233 and uranium 235, d. calcining in an argon atmosphere ata temperature of about 1, 000* to 1,900*C. for about 3.5 to 30 hours,followed by e. calcining in a nitrogen atmosphere at a temperature of1,000 to 1,900C. for about 6 to 10 hours, and f. recovering the mixednitride nuclear fuel product.
 7. The process according to claim 6wherein the impregnated particles contain about 1 to 30 percentfissionable materials.
 8. The process according to claim 6 wherein theparticles are calcined in each of the calcination steps for a period ofabout 24 hours.
 9. The process according to claim 5 wherein the urania238 particles are heated in hydrogen or other reducing atmosphere at atemperature of about 400*-700*C. for a period of 0.5 to 3 hours.
 10. Theprocess according to claim 6 wherein the urania 238 particles are heatedin hydrogen or other reducing atmosphere at a temperature of about400*-700*C. for a period of 0.5 to 3 hours.